ML18005A667

+R. Biggerstaff, Principal Engineer,

Onsite Nuclear Safety

C. Crawford, Director, Onsite Nuclear Safety

D. Elkins, Radioactive

Waste

Foreman,

Radiation Control

J. Floyd, Operations

Foreman Radiation Control

"+C. Gibson, Director, Programs

and Procedures

"+C. Hinnant, Plant General

Manager

J. Leonard, Specialist,

Radioactive

Waste

+J. McDuffie, Supervisor,

Radiation Control

+T. Morton, Manager,

Maintenance

J. O'Halloran,

Foreman,

Dosimetry

C. Olexik, Supervisor,

Operations

+A. Poland,

Project Specialist,

Radiation Control

+M. Pugh, Project Specialist,

In-service Inspection

F. Reck,

Foreman,

Support

+J. Sipp, Manager,

Environmental

and Radiation Control

D. Tibbitts, Director, Regulatory Compliance

+R.

Van Metre, Manager,

Technical

Support

M. Wallace, Senior Specialist,

Regulatory Compliance

R. Watson,

Vice President,

Harris Nuclear Plant

L. Williams, Principal Engineer,

Corporate

Other

licensee

employees

contacted

during this

inspection

included

craftsmen,

engineers,

technicians,

and administrative personnel.

Other Organizations

Nuclear Regulatory

Commission

(NRC), Region II

P. Fredrickson,

Section Chief, Division of Reactor Projects

C. Hehl, Deputy Director, Division of Reactor Projects

C. Hughey, Radiation Specialist,

Division of Radiation Safety

and

Safeguards

E. Merschoff, Deputy Director, Division of Reactor Safety

NRC Resident

Inspectors

+W. Bradford, Senior Resident

Inspector

G. Maxwell, Senior Resident Inspector

wM. Shannon,

Resident

Inspector

Attended exit interview on August 19,

1988

+Attended exit interview on September

2,

1988

Acronyms

and initialisms

used

throughout this report are listed in the

last paragraph.

2.

Occupational

Exposure

During. Extended

Outages

(83729)

a ~

Organization

and management

Controls

(83722)

Organization

The licensee

is required

by TS 6.2 to establish specific onsite

and offsite organizations

for unit operation

and

corporate

management.

The responsibility, authority and other management

controls

necessary

for establishing

and maintaining

a health

physics

program for the facility are further outlined

in

Chapters

12

and

13 of the

FSAR.

TS 6.5.2

specifies

the

composition of the Plant

Nuclear Safety

Committee

(PNSC)

and

delineates its functions, responsibilities

and authority.

The inspector

reviewed the licensee's

site organization,

as well

as

the

responsibilities,

authority

and

control

given

to

management

as

they relate

to the site radiation protection

program.

Recent

changes

in plant management

were reviewed to

verify that they would not adversely affect the'bility of the

licensee

to continue

implementing the critical elements

of the

program.

The inspector

discussed

the support

received for the

radiation

protection

program

with the

ESRC

Manager

and

determined that it was adequate.

(2)

Staffing

TS 6.2 also specifies

the minimum staffing requirements for the

facility.

FSAR Chapters

12

and

13 outline further details

on

staffing levels at the site.

The inspector

reviewed the

Rad

Con or

HP organization with the

ESRC Hanager.

The subjects of attrition rate,

use of contractor

HP personnel,

promotions, staff qualifications

and actual

versus

authorized staffing levels were also reviewed

and discussed.

At

the time of the inspection,

22 of the

27 authorized

senior or

ANSI qualified technician positions

were filled.

The licensee

indicated that there

was

an active recruitment

program in place

and that the five empty positions

would be filled as

soon

as

personnel

accepted

job offers.

It was

also

noted that

the

licensee

had

12 junior technicians

in training status.

The

licensee

indicated that the junior technicians

were being

used

during the current outage

to perform closely supervised

routine

jobs.

Licensee

representatives

also indicated that they would

be qualified by February

1989.

In addition

to the

permanent

personnel,

the

licensee

had

acquired

the

assistance

of

70 contractor

HP technicians

to

augment

the staff during the outage.

Upon completion of the

outage,

the licensee

planned to reassess

the personnel

needs

in

the

group

and explore

the possibility of increasing

the

"permanent

staff,

especially

in

the

ALARA section

(see

Paragraph f(1) ) .

{3)

Management

Controls

The inspector

reviewed the licensee's

Radiation Safety Violation

(RSV) reports

which were

used to identify and document safety

and radiological incidents.

It was noted that most of the

RSVs

involved

personnel

failure to follow good radiological

work

practices

or failure to comply with instructions.

The inspector

discussed

these

problems

with the

ESRC Manager

who indicated

that these

problems would likely be eliminated

as the work force

became

more

experienced

in dealing with actual

radiation

and

contamination.

It was

noted that

adequate

corrective actions

had

been initiated, as required.

No violations or deviations

were identified.

b.

Training and gualifications

(83723)

10 CFR 19. 12 requires that all individuals working in or frequenting

any portion of a restricted

area shall

be provided basic radiation

protection training.

The licensee's

General

Employee Training

(GET) was divided into two

parts.

If individuals were only entering the'protected

area

then

Level I training was required.

Level I required approximately

seven

hours of classroom

work with

a written test

and

covered

workers

rights,

conduct,

and plant specifics.

If the individual needed

to

enter the radiation control area,

Level II GET was required.

Level

II required approximately nine hours of classroom

work which included

a test

and "mock-up".

"Mock-up" was the practical factors portion of

the course

and required the worker to demonstrate

the proper use of a

survey

instrument,

proper

dress-out

in protective

clothing,

RMP

comprehension,

and

SRPD reading.

The inspector

reviewed lesson

plans

for Level I and !I and verified that all topics specified in 10 CFR 19. 12 were covered during the training.

Selected training records of

licensee

and contractor

personnel

working in the radiation control

area

were

reviewed.

Licensee

representatives

stated

that

GET

retraining

was

given

every

twelve

months

and

consisted

of

approximately four and

a half hours of classroom work, "mock-up", and

a written test.

Respirator

training,

which required

approximately four hours

and

included

hands

on work, was reviewed

by the inspector

and found to be

adequate.

Training personnel

stated,

that since the staff could be supplemented

with trainers

from other training groups,

an adequate

number of GET

trainers

were available for the

increased

number of individuals

requiring

GET training due to the outage.

Because

the inspector

observed

individuals improperly wearing

PCs

within the

RCA (see

Paragraph 2.e.(3)), specific attention

was given

to the type and amount of information conveyed to the worker relating

to

PC donning

and wearing.

Training appeared

adequate

in this area;

however,

more emphasis

may be required.

Through discussions

with training personnel

the inspector determined

that

a good line of communication existed

between operational

HP and

GET training personnel

to quickly address

any possible

poor work

practices identified in the field through improvements in training.

No violations or deviations

were identified.

External

Exposure Control

and Dosimetry

(1)

Dosimetry

The licensee is required

by 10 CFR 20.101

and 20.102 to maintain

workers'oses

below specified levels.

10 CFR 20.202 requires

each licensee

to supply appropriate

personnel

monitoring devices

to specific individuals and require the use of such equipment.

The licensee's

external

exposure

control

and personnel

dosimetry

programs

were

reviewed

by

the

inspector.

This

included

facilities, equipment,

personnel,

records,

and procedures

used

to control exposures

and determine

doses.

The

licensee

uses

Panasonic

four element

thermoluminescent

dosimeters

(TLDs) which are routinely read

on

a quarterly basis.

The Dosimetry

Program is due for National Voluntary Laboratory

Accreditation

Program

(NVLAP) accreditation

renewal

October 1,

1988.

The licensee

also participates

in a

TLD spiking program

with Battelle

National

Laboratory.

Licensee

representatives

stated

that

they will soon

begin

using

a

newly developed

finger-ring dosimeter

from Panasonic for extremity 'monitoring of

the

hands.

Previously,

the licensee

used

band-aid

type finger

dosimeters

with Panasonic

elements.

(2)

Administrative Control Levels

The inspector

reviewed

the licensee's

administrative

exposure

controls

and

determined

that

those

controls

were designed

to

maintain exposure

as

low as reasonably

achievable

(ALARA).

The

licensee

required

consecutively

higher tiers of supervision to

approve

dose

extensions

with the plant manager

having approval

5

authority for whole body exposures

in excess

of 2400 mrem during

the calendar quarter.

Records

Exposure

records

of plant

and contractor

personnel

for 1988,

year-to-date,

were selectively

reviewed.

No exposure

greater

than limits in

10 CFR 20.101

or the station's

quarterly

administrative control level

was noted.

A review of. the station

Dose

Report,

dated

September

1,

1988,

indicated

a

maximum

individual dose of 924 mrem for the current quarter.

Skin and

extremity doses

were negligible.

Steam Generator

(S/G)

Work

The inspector

reviewed the licensee's

procedure for working in

and

around

steam

generators,

Health Physics

Procedure

HPP-010,

Steam

Generator

Entry, Revision 3, dated

August 22,

1988.

It

was noted that the procedure

contained

references

to protective

clothing,

respiratory

protection,

dosimetry

requirements

and

exposure

control,

as well

as

"jump sheets"

to control entries.

The

Procedure

also

specified

requirements

for radiological

surveys

which were

performed

by the licensee

on August 28-29,

1988,

during

and following the

removal of the

manway covers,

diaphragm

and inserts for the three

S/Gs.

The results of the

surveys

indicated

that

beta

radiation levels just inside the

manway

openings

were from 4 to

5 times higher than the

gamma

radiation levels,

the highest

being

44

Rad per hour

beta

and

10 R/hr

gamma.

The inspector

discussed

beta

radiation

exposure

control with

licensee

representatives

and

reviewed

the

licensee's

investigation

of the

isotopes

present

in the

S/Gs

and

the

energies

involved.

This

was

being

done to ensure

that the

controls

being

used

during

S/G entries

were

adequate

and to

assess

the

possible

need for other

measures

such

as

beta

radiation

stay

times.

It was

noted

that,

although

the

licensee's

procedure

required

surveys

prior to

S/G entry,

no

mention

was

made

of the

beta

radiation

exposure

control

evaluations.

The

licensee

acknowledged

this

and

agreed

to

revise the procedure

to add

a requirements for such evaluations

prior to

S/G entry.

In addition,

the

licensee

agreed

to

consider

inclusion of attentuation

studies

using various

samples

of clothing and eye protection in the evaluations

to be required

by procedure.

The

inspector

reviewed

the

TLD dose

value results

from the

multiple badge

sets

worn by workers during various

S/G entries,

These results

did not indicate

any lens of the eye dose greater

than the whole body dose.

High Radiation Area Control

Technical

Specification

6. 12

requires

that

any

individual

permitted to enter

a high radiation area in which the intensity

of radiation is equal

to or less

than

1000 millirem per hour at

'8

inches shall

be provided with or accompanied

by:

(a)

a radiation monitoring device that continuously indicates

the radiation

dose rate in the area;

or

(b)

a radiation monitoring device that continuously integrates

the radiation

dose

rate

in the

area

and

alarms

when

a

preset integrated

dose is received;

or

(c)

an individual qualified in radiation protection procedures,

with a radiation dose rate monitoring device.

During tours of the Unit 1 containment,

the inspector verified

dose

rates

in various

areas,

throughout

containment

and

high

radiation areas

(HRAs).

The inspector also reviewed survey

maps

of various

areas

in the

RCB and

noted that

dose

rates

up to

800 mrem/hr at contact

and

250 mrem/hr at

18 inches existed in

the area

near the

"A" S/G inside the biological shield.

Other

areas

within the containment

were noted with similar radiation

levels.

During tours

on August 15-17,

1988, the inspector

observed

work

in progress

on all elevations

in the

RCB including work in HRAs.

During that time period the inspector

noted that the licensee

issued

a survey meter or

an integrating/alarming

dosimeter to

individuals entering

HRAs or to groups of individuals entering

HRAs who would be working in the

same location.

The inspector

questioned

three individuals

on August

16 and five individuals

on August

17 about the use of these

devices to monitor radiation

dose

rates

and control

exposure

and found that all individuals

either

had

a meter or an alarming dosimeter or were accompanied

by someone

who had one.

On the morning of August 18,

1988, at approximately 9:30 a.m.,

a

worker was

observed within a posted

HRA inside the biological

shield in Unit 1

RCB.

The inspector

questioned

the individual

as to the nature of the work he

was performing

and about the

possession

and use of an integrating dosimeter or a survey meter

in a

HRA.

The worker indicated that

he was performing snubber

inspections

under the requirements

of

SRWP 318,

Inspection

and

Functional

Testing

of

$nubbers,

Scaffolding

Support

and

Support,

which required individuals to have access

to all areas

inside

and

outside

the biological shield.

When

asked,

the

individual

stated

that

he

had

neither

a

meter

nor

an

integrating/alarming

dosimeter.

The worker left the

HRA and

went to the containment

control point and obtained

an alarming

dosimeter for work in the

HRA.

The inspector also reviewed the

requirements

of

SRWP 318 which required

the

use of

a survey

meter (or alarming dosimeter)

in HRAs.

Failure of the licensee

to provide the worker with a survey

meter or an integrating/alarming

dosimeter or to have the worker

accompanied

by an individual qualified in radiation protection

procedures

with

a monitoring device for entry into

a

HRA was

identified

as

an

apparent

violation

of

TS 6.12

(50-400/88-28-01)-.

d.

Internal

Exposure Control

and Assessment

(83725)

Engineering

Control s

10 CFR 20.103(b)(1)

requires

that the licensee

use

process

or

other engineering

controls to the extent practicable

to limit

concentrations

of radioactive

materials

in the air to levels

below those

which delimit an airborne radioactivity area

as

defined in 20.203(d)(1)(ii).

During tours of the Auxiliary Building and the Unit 1 Reactor

Containment Building, the inspector

observed

the use of various

engineering

controls

employed

to limit the concentrations

of

radioactive material

in air.

The licensee

used

enclosures

or

tents

constructed

around

equipment

or

areas

which were highly

contaminated

and provided

HEPA filtered ventilation systems for

the enclosures.

In other areas,

the licensee

used ducting to

draw air

away

from contaminated

areas

and into the filtered

ventilation system.

Nachining or grinding was also performed in

tents

or using

directed

ventilation to control

and limit

airborne radioactivity.

The inspector also observed

the use of

a continuously oscillating water sprayer

used to keep the upper

internals of the reactor wet while stored in the drained portion

of the cavity.

(2)

Respiratory Protection

Program

10 CFR 20.103(b)

requires

that,

when it is impracticable

to

apply process

or engineering controls to limit concentrations

of

radioactive

material

in air below

25% of the concentrations

specified in Appendix B, Table 1,

Column 1, other precautionary

measures

should

be

used

to maintain the intake of radioactive

material

by any individual within seven

consecutive

days

as far

below 40 Naximum Permissible

Concentration-hours

(NPC-hrs)

as is

reasonably

achievable.

Through

records

review,

observations

and

discussions

with

licensee

representatives,

the

inspector

evaluated

the

respiratory

protection

program

including training,

medical

qualifications, fit testing,

NPC-hr assignment,

and issue,

use,

(3)

(4)

decontamination

and storage

of respiratory protection devices.

The records,

observations

and discussions

indicated that only

those

personnel

who have

been trained

and qualified to wear

a

respiratory protective

device

were issued respirators.

Review

of the MPC-hr assignments

for selected

individuals revealed that

all exposures

were

below the

40 NPC-hr per week control level

and below the facility's administrative control levels

as well.

Air Sampling

and Bioassays

10 CFR 20. 103 establishes

the limits for exposure of individuals

to concentrations

of radioactive materials in air in restricted

areas.

Section 20.103 also requires that suitable

measurements

of concentrations

of radioactive material in air be performed to

detect

and evaluate

the airborne radioactivity in restricted

areas

and that appropriate

bioassays

be performed to detect

and

assess

individual intakes of radioactivity.

The inspector

reviewed the results of selected air samples

taken

during the current outage.

The air sample

log indicated that

the airborne concentration

had

seldom

been

above

25K of the

NPC

of radionuclides

specified in 10 CFR 20,

Appendix B, Table 1,

Column 1.

The licensee

also indicated that

no problems

had

been

encountered

during the

outage,

to date,

concerning

airborne

radioactivity including radioiodine.

It was noted that the air

samples

had

been

evaluated for alpha,

beta

and

gamma activity

and analyzed to determine the specific isotopes

present.

The results of selected

WBCs were also reviewed.

No instances

were

noted in which personnel

received greater

than the limits

specified

in

10 CFR 20.103.

Through

records

review

and

discussions

with licensee

representatives,

it was

determined

that

there

had

been

no detectable

intakes

of radioactive

material

since

the last inspection

and

no confirmed internal

exposures.

Air equality

The

inspector

discussed

with licensee

representatives

their

procedures

for

ensuring

air quality

supplied

to airline

respirators,

bubble

hoods

and

self-contained

breathing

apparatuses

(SCBAs).

It was

noted

that

a

dedicated

air

compressor

was

used

to fill air bottles

used

in SCBAs.

The

compressor

was tested

every

3 months to maintain

Grade

D air.

Plant instrument air was used for airline respirators

and bubble

hood air supply

and air quality was

checked

onsite

every

3

months or the first time

a connection

was made to

a given outlet

to ensure

Grade

D air was always available.

No violations or deviations

were identified.

Control

of Radioactive

Material

and

Contamination,

Surveys

and

Monitoring (83726)

Contamination

Surveys

10 CFR 20.201(b)

requires

each licensee

to make or cause

to be

made

such

surveys

as

(1) may

be necessary

for the licensee

to

comply with the regulations

in this part and (2) are reasonable

under

the circumstances

to evaluate

the extent of radiation

hazards that may be present.

Health

Physics

Procedure

HPP-030,

Control

and

Release

of

Equipment/Material

from the

RCA/Restricted

Area,

Revision 3,

dated

August 27,

1987,

requires

in

Section 10.1.1

that

material/equipment

being transferred within the

RCA be surveyed

for radiation

and

contamination

per

HPP-060,

Performance

of

Radiation

and

Contamination

Surveys.

Also,

Section 10.1.2

states

that, if radiation/contamination

levels

are

less

than

100 net

cpm per

probe

area

as measurable

within a thin window

pancake-type

GM detector

by direct frisk and

less

than

MDA

(defined

as

100 net

cpm per

100 cm~) for the counting instrument

used

by smear

survey (removable),

the material/equipment

can

be

released

for transfer.

Health Physics

Pro'cedure'PP-060,

Performance

of Radiation

and

Contamination

Surveys,

Revision 2,

dated

September

29,

1987,

requires

in

Section 7.12

that,

when

counting

smears

for

beta-gamma

contamination

using

a frisker,

a minimum count time

of about

20 seconds

is

recommended

to allow for a

90 percent

deflection of full scale

on the

slow response

mode.

Also,

Section 10.8.2 requires

a frisk of the accessible

surfaces

of

the material

being surveyed

by moving the probe slowly over the

surfaces within one-half inch of the surface.

While touring the Auxiliary Building, the

RCB and the

Radwaste

Building

to

determine

whether

or

not

adequate

personal

contamination

surveys

were

being

performed,

the

inspector

observed

workers exiting contaminated

areas

and exiting the

RCA.

The

movement of material

from contaminated

areas

and from the

RCA .also

was

observed

to determine if adequate

direct

and

smearable

contamination

surveys

were

being

performed.

All

personnel

surveys

observed

appeared

to

be

adequate

and

the

material

surveys

performed

at the exit of the

RCA were also

adequate.

However, the contamination

surveys

performed at the

main

control

point to

the

RCB, for items

cross'ing

the

contaminated

area

boundary,

were less

than adequate.

On

numerous

occasions

during the period of August 15-18,

at

varying times during the day, the inspector

noted

HP technicians

surveying

items out of the

RCB contaminated

area

at the

main

10

control point.

Various survey techniques

were observed

during

the time period,

the most prevalent of which was the

use of a

white cotton liner to "smear"

an item. If the liner was worn on

the hand,

the fingers of the liner would be wiped over

a portion

of the item being .surveyed

and then checked for contamination

by

placing

the, fingertips over

a frisker probe for 3-5 seconds.

Another method involved the

use of a cotton liner to wipe over

the item in question

continually folding the liner as it was

used.

The liner would then

be

checked for the presence

of

contamination

by placing the folded liner over

a frisker probe

for 3-5 seconds.

The liner was. not unfolded or spread

out on a

flat surface

to frisk the entire surface

used to "smear" .the

item.

When

a

smear

patch

was

used

to

smear

an item, it was-

seldom

counted for more than

2 seconds

with a frisker probe.

Only

on infrequent

occasions

were

items

such

as electrical

cords,, hoses,

cables

or other pieces of equipment wiped down or

smeared

along their entire length or over the entire

surface

area to check for contaminat'ion.

(3)

The inspector also noted that such items were not frisked with a

probe

along their entire length or over their entire external

surface.

Usually only a cursory frisk of the item was performed

before

removing

the

item from the

contaminated

side of the

barrier

and

placing

the

item

on the

"clean"

side of the

boundary.

Failure to perform

an

adequate

survey of items being released

from a contaminated

area

was identified as

an apparent violation

of 10 CFR 20.201(b)

(50-400/88-28-02).

Personnel

Contamination

Reports

The inspector

reviewed selected

personnel

contamination reports

for 1988

and all reports

of contamination

that

had occurred

since

the current

outage

began.

It was

noted that, of those

involving skin contamination,

most were determined to have

been

caused

by tom

PCs

or from improperly

removing

the

PCs.

Licensee

representatives

indicated

an

awareness

of these

problems

and

had

made

suggestions

and

recommendations

to the

training group about the proper

use

and

removal

of PCs.

The

inspector verified that the training group

had

been

advised of

these

issues

and

had

taken

steps

to emphasize

the correct

use

and removal of PCs in GET and retraining. classes.

Protective Clothing (PCs)

During tours of the

RCB, the inspector

observed

various jobs in

progress

including installation

of the

guide

pins

in the

reactor,

snubber

inspection,

installation

of

shielding,

decontamination efforts following a spill from the

RTD manifold,

as well

as other maintenance

activities.

During these

tours,

the inspector

also

noted isolated

instances

in which workers

were wearing the

PC cloth hood with the flaps tied or attached

behind their

heads.

Two workers were also

noted to be wearing

their

PC coveralls with the front zipper partially undone

thus

exposing

the upper portion of the neck

and chest.

When asked

about this practice,

licensee

HP representatives

indicated that

this was

an improper practice

and that this would be monitored

during future

management

and

HP tours

of the

containment.

Through review of personnel

contamination reports,

the inspector

found

no

evidence

that this

problem

was contributing to

an

increased

incidence

of personnel

contaminations.

This issue

will be

reviewed

during

subsequent

inspections

and will be

tracked

by the

NRC

as

an

Inspector

Followup

Item (IFI)

(50-400/88-28-03).

Radioactive Material Labeling and Storage

10 CFR 20.203(e)

requires

each

area

or room in which licensed

material

is

used

or stored

and which contains

any radioactive

material

in an

amount

exceeding

10 times the quantity of such

material

specified

in

Appendix

C of this

part

shall

be

conspicuously

posted with a sign or signs bearing the radiation

caution

symbol

and the words:

"CAUTION, RADIOACTIVE MATERIAL."

10 CFR 20.203(f)

requires

that

each

container

of radioactive

material

bear

a durable, clearly visible label identifying the

radioactive contents.

During tours of the

RCA, adjacent

storage

areas

and warehouses,

the

inspector

observed

the

licensee's

radioactive

material

storage

areas.

The

rooms

or

areas

used

for storage

of

r'adioactive material

were posted

as required

and the items kept

therein

were properly labeled.

It was noted that the licensee

had very little radioactive material in storage.

Radi ati on Detecti on and Survey Instruments

During plant tours,

the inspector

observed

the proper

use

and

selection

of

instruments

appropriate

for the

radiation

protection activity in progress.

The inspector verified that

instruments

in use

or available for use

had

been calibrated

within the prescribed

time period.

Licensee

personnel

stated

that

the quantity

and

type of portable radiation detection

instruments

were adequate for the increased

radiation protection

activities resulting from the outage.

The

inspector

discussed,

with

licensee

representatives,

operation

and calibration

of the

whole

body

contamination

monitors

permanently

located at the

RCA exit and,

during the

outage,

at

the

main

containment

control

point.

Licensee

representatives

stated that three

Nuclear Enterprise

IPM-7s and

12

six Helgeson

2As were currently onsite.

The inspector

reviewed

procedures

SIC-046,

Calibration of

IPM-7, Revision 1,

dated

November 25, 1986,

and SIC-045, Calibration of HECM-2A, Revision

0,

dated

May 19,

1987,

and

inquired

as

to the

supporting

documentation

for the

set-up

parameters

dictated

by these

procedures.

Licensee

representatives

stated that the vendor's

Technical

Manuals

were

used

and

supplemented

by start-up

operational

testing

performed

by the licensee.

The licensee

could not supply specifics of the start-up testing

because

the

reports

documenting

the tests

had been lost.

Concerns

raised

by

the inspector,

such

'as

the monitors'etection

efficiency for

commonly

found

isotopes

other

than

the

Cs-137

used

for

calibration

and the difference in the monitors'ensitivity for

point sources

versus

plate

sources,

were not addressed

in the

Technical

Manuals.

The inspector

discussed

with the licensee

the adequacy of personnel

contamination

surveys

and the need to

review the documentation

of the operational

tests

conducted

by

the licensee.

The licensee

agreed that documentation of the old

operational

tests

should

be obtained

or that

new operational

tests

be

performed

and

properly

documented.

Documentation

supporting

the set-up

parameters

used for'he

IPM-7 and the

HECM-2A personnel

contamination monitors will be reviewed during

a

subsequent

inspection

and

tracked

by the

NRC as

Inspector

Followup Item (IFI 50-400/88-28-04).

Response

Checks

TS 6.12 allows individuals to enter high radiation areas if they

are

provided

with

a

radiation

monitoring

device

that

continuously integrates

the radiation

dose rate

and alarms at

a

preset integrated

dose.

TS

6. 11. 1 requires

that

procedures

for personnel

radiation

protection shall

be prepared

consistent with the requirements

of

10 CFR 20 and shall

be approved,

maintained

and adhered

to for

all operations

involving personnel

radiation exposure.

Health Physics

Procedure

HPP-460,

Operation of the Dositec

502A,

Revision 1, dated

June

25,

1987, states

that

a Dositec

502A may

be used if specified

acceptance

criteria are met,

one of which

is passing

a response

check.

The inspector

discussed

the source

check

and response

check of

Dositecs

with

licensee

representatives

and

reviewed

documentation

of the results.

Through discussions

and record

reviews,

the

inspector

verified that

during

the

period of

August 15-17,

1988, all

Dositec

502As

issued

from the

main

control point of the

RCP were not. response

checked prior to the

instruments

being

used.

The inspector

informed the licensee

that failure to perform the required

response

checks

was

an

apparent violation of TS 6.11. 1 (50-400/88-28-05).

13

f.

Program for Maintaining Exposures

As

Low As Reasonably

Achievable

(ALARA) (83728)

10 CFR 20. 1(c) specifies

that licensee

should

implement

programs

to

maintain workers'oses

ALARA.

Other

recommended

elements

of an

ALARA program are outlined in Regulatory

Guides 8.8 and 8.10.

The

FSAR, Chapter

12, also contains

licensee

commitments

regarding worker

ALARA actions.

(1)

Site

ALARA Group

Prior to the

outage,

the

ALARA group

was

composed

of one

specialist

and

one contract technician.

During the outage,

the

ALARA group

was

augmented

by

two

additional

contract

technicians.

Their major functions

were

to

review Plant

Operating

Manual

procedures,

perform pre-plan

and

post-job

reviews,

provide

trend

analyses

of

such

items

as

current

personnel

exposure,

job exposure,

and personnel

contaminations

and establis'h

and track trends

against established facility and

group exposure

goals.

The ALARA personnel

were also required to

make tours of the plant

and observe

jobs in progress

whenever

possible.

The inspector discussed

the staffing of the

ALARA group with the

E&RC Manager.

While the group

was able to perform the various

tasks

assigned, it was

apparent

that

more

personnel

would

facilitate better tracking of exposures

and more field coverage.

Licensee

management

indicated that the subject of increasing

the

size of the

ALARA staff had been considered

and that, following

the

outage,

an

increase

would again

be evaluated

and

given

serious consideration.

(2)

ALARA Subcommittee

The

ALARA Subcommittee

was

composed of representatives

from each

of the major work groups onsite.

The

PNSC representative

on the

subcommittee

was the

E&RC Manager with the

ALARA specialist

from

the

group

acting

as

committee

chairman.

The

group

was

established

to report to the

PNSC

and

make

recommendations

to

management

on

ways

to

maintain

exposures

ALARA.

The

subcommittee

also

functioned

as

a review committee for jobs

whose total

exposure

exceeded

25 person-rem.

The majority of

the

recommendations

made to date

consisted

of suggestions

for

plant modifications.

The committee

chairman also tracked these

recommendations

and the progress

made in implementing

them or

lack thereof.

The

subcommittee

meets

monthly

as

outlined

by procedure

to

review past

performance

and

recommendations

and

consider

new

recommendations.

The inspector

reviewed

the minutes

of the

subcommittee

meetings for 1988

and the recommendations

that had

14

been

made.

It was noted that

many of the recommendations

were

still pending.

The licensee

indicated that,

because

most of the

recommendations

dealt with plant modifications,

a great deal

of,

time

was often required

to implement the

ones that

had

been

adopted.

Job Review

The inspector discussed

the job evaluation

and review process

in

detail

with the

ALARA specialist.

Prior to

a job being

performed,

the job coordinator

was required to determine if an

ALARA prejob review was required.

If a review was required,

the

job coordinator filled out

a checklist to ensure that various

aspects

of the

job were

or

had .been

considered,

such

as,.

approval

of

procedures,

staging

of

necessary

materials,

availablity of service 'lines,

and availability of communications

equipment.

The

ALARA specialist

or his

designee

would then

complete

a checklist

addressing

such

items

as

review of

historical

data,

temporary

shielding,

and

decontamination

requirements.

For

high risk

or

complicated

jobs,

a prejob

review would be held including representatives

from the various

groups

involved to perform

and cover the work.

During such

a

review,

the

actual

details

of the job evolution would

be

reviewed

and the expected

actions -of the workers

and the results

of the work discussed.

The

inspector

also

reviewed

the job evaluation

and

review

procedure,

Administrative

Procedure

AP-514,

Job

Evaluations,

Revision 1,

dated

December

12,

1986.

While the

procedure

specifically

covered

pre-plan

and

post-job

reviews, it did not address

the subject of review of jobs/work

in progress.

The licensee

indicated that,

when the scope of the

job changed

which could entail

more exposure,

the

HP technician

covering

the

work would inform the

ALARA group.

The

specialist

would then review the

change

and make changes

to the

exposures

estimated

as

necessary.

Licensee

representatives

indicated

that

the

procedure

was

under

review

and

would

be

revised,

following the outage,

to facilitate its usage

in the

future.

The inspector

reviewed

numerous

ALARA pre-job reviews

and the

associated

documentation

delineating the estimated

person-rem to

be

expended,

the aspects

of the work considered,

any extra or

additional

requirements

established

as

a result,

and

the

required approvals.

The reviews included

S/G opening

and entry,

weld inspections,

eddy current testing,

and sludge lancing.

The-

estimates,

work reviews

and additional

requirements

appeared

to

be adequate.

All pre-job reviews

had been

approved

as required

by procedure.

15

(4)

Job Performance

The

inspector

reviewed

the

jobs

being

performed

during the

outage.

The licensee

indicated that,

as of September

1,

1988,

they were in day

34 of

a scheduled

56-day outage.

The major

jobs

performed

and the total

dose

received

in person-rem for

each job were:

Job

Pe'rson-rem

HP Coverage

5 Surveys

Scaffold

5 Insulation Work

Guide Stud Repairs

S/G Nanway Removal

Nozzle

Dam Installation

Decontamination

Section

XI Weld Inspection

S/G Secondary

Work

Reactor

Head Work/Refueling

Reactor Coolant

Pump

PNs

Snubber Inspections

Routine Naintenance

Nanagement

Inspections

(5)

Annual

and Outage

Exposure

Goals

13.2

12.1

1.6

4.1

8.5

3.4

10.3

6.3

8.3

3.1

4.6

2.6

The person-rem

exposure

goal for the outage

was set at 200 with

an

annual

goal of 363 person-rem.

As of August ll, 1988, the

licensee

had

expended

approximately

63 person-rem,

including

both routine

and outage activities,

as determined

by SRPD.

The

licensee

had only expended

33 person-rem during their first year

of operations

in

1987.

Personnel

contaminations

for

1988

totaled

67, with

17 of those

occurring during the outage

to

date.

The

licensee

experienced

a total

of

87

personnel

contaminations

during

1987,

48

skin

and

39

clothing

contaminations.

No violations or deviations

were identified.

3.

Solid Waste

(87422)

10 CFR 20.311 requires

a generating

licensee

who transfers

radioactive

waste

to

a land disposal facility or to

a licensed

waste collector to

prepare

all waste

so that the waste is classified

according

to

10 CFR 61.55

and

meets

waste characteristic

requirements

of

10 CFR 61.56.

The

regulation

further establishes

specific

requirements

for conducting

a

quality control

program.

16

Waste Characterization

and Classification

Licensee

representatives

stated that updated

scaling factors

used to

classify waste

streams

at the plant were received

from an offsite

vendor the

week of the inspection.

The frequency of these

updates

had

previously

been

on

a

quarterly

cycle;

however,

licensee

representatives

stated

that

beginning

next quarter,

waste

stream

sampling

and

analysis will be

on

an

annual

frequency.

Special

samples will be taken if anomalies in a system which is the source of

a waste

stream

become apparent.

Shipments

As of August 17,

1988, eight shipments

of evaporator

concentrates

solidified in

cement

had

been

made

to

a disposal

site

and

one

shipment of dry active

waste

(DAW)"was transferred

to

a licensed

waste collector.

(vendor supercompactor).

- The inspector

rev'iewed the

waste manifest for these

shipments

and verified that

10 CFR 20.311

requirements

had

been

met.

Licensee

representatives

stated that

no

dewatered

resins

had

been

shipped this year.

The inspector

noted

that the licensee

has

never

shipped

Chemical

Volume Control

System

(CVCS) resins

or primary water

system filters both of which have

higher specific activities

than material

currently

being

shipped.

However,

CYCS resins

and primary filters will be processed

in the

near

future

and

subsequently

shipped.

Licensee

representatives

stated

that

as

of July 20,

1988,

5,330 ft~ of radioactive

waste

containing

1.825 curies of activity had

been shipped offsite and 824

ft~ was stored onsite awaiting shipment.

Waste Segregation

The inspector

discussed

the methods

and procedures

used for release

of. "clean"

trash

and

segregation

of "contaminated"

trash

with

licensee

representatives

and

reviewed

licensee

procedure

HPP-100,

Segregation

and

Release

of

Waste

from the

RCA/Protected

Area,

Revision 2, dated

December

15,

1987.

All trash

taken from the

RCA,

both clean

and potentially contaminated,

is spread

out under

a

HEPA

filtered hood.

Each

piece is then

surveyed

and segregated.

The

"clean"

bags of trash

are

then

taken to

a low background

area

and

surveyed with a microR meter

and released

to the licensee-controlled

landfill.

Licensee

representatives

stated that the landfill is also

surveyed

with

a microR meter

once

a week.

Multiple surveys of the

trash

prior to its

removal

from the

RCA is indicative of

a

radioactive

waste

program

committed

to ensuring

that radioactive

contamination is not released

to uncontrolled areas.

Process

Control

Program Procedures

TS 6.8.1 requires

that procedures

be established,

implemented,

and

maintained

covering Process

Control Program'(PCP)

implementation.

TS 6.8.2

requires

that

each

procedure

of Specification

6.8.1,

and

17

changes

thereto,

shall

be reviewed

and

approved

in accordance

with

Specification 6.5. 1 prior to implementation.

During a review of the vendor supplied procedures

which implement the

Program for solidification of radioactive

waste,

the inspector

determined

that

no review

and

approval

conforming with Technical Specification 6.5.1

was

documented.

Licensee

representatives

stated

that

an informal review was conducted;

however, this review was not

documented

and

no

formal

approval

was

obtained

prior to the

initiation of waste solidification and

implementation

of the

procedures.

The licensee

produced

documentation

showing that the -NRC

had accepted

the vendor's topical report describing the mobile cement

solidification system

and

documentation

showing that

NRC acceptance

is pending for the vendor's

topical report which demonstrates

that

the

vendor's

final solidified product

meets

waste

structural

stability requirements

specified in 10 CFR Part 61.

The inspector

stated

that

these

NRC approvals

do not address

implementing

procedures

and that failure to review and approve

the

implementing

procedures

prior to implementation

was

an apparent

violation of TS 6.8.2 (50-400/88-28-06).

4.

Transportation of Radioactive Materials

(86721)

I

a.

Shipping Records

10 CFR 71.5(a)

requires

that

each

licensee

who transports

licensed

material

outside of the confines of its plant or other place of use,

or who delivers licensed material

to

a carrier for transport,

shall

comply

with

the

applicable

requirements

of

the

regulations

appropriate

to

the

mode

of transport

of the

Department

of

Transportation

(DOT) in 49

CFR Parts

170 through 189.

In addition

to the

shipments

referenced

in Paragraph

3.b.,

the

inspector

reviewed selected

"special" radioactive material

shipments

shipped

in

1988.

These

shipments

included

such

items

as

hot

particles,

waste

stream

samples,

and

an alloy analyzer.

All

transportation

documentation

reviewed complied with applicable

49

CFR

regulations.

b.

Site Transportation

Organization

The transportation/shipping

staff was

composed of a foreman,

who also

assisted

HP operations

during the outage,

two permanent technicians,

and

a third position which was filled with

HP technicians

rotated

from the

HP operations

group.

Specialized training for the staff was

adequate

with retraining given at an appropriate

frequency.

18

c.

guality Assurance

Audits

Licensee

representatives

stated

that

a guality Assurance

audit

completed

in July 1988,

covered

proper

shipment

packaging

and

shipping

paper completion.

The report had not been

completed at the

time of the inspection

and

was not reviewed

by the inspector.

The

audit

was

conducted

by

gA personnel

supplemented

with technical

specialists

from other licensee facilities.

Licensee representatives

stated that no findings were identified in the area of transportation

and shipping.

No violations or deviations

were identified.

5.

Exit Interview

The inspection

scope

and results

were summarized

on August 19,

1988 and

on

September

2,

1988,

with those

persons

indicated

in Paragraph

1.

The

inspector

described

the

areas

inspected

and

discussed

in detail

the

inspection findings listed below.

No dissenting

comments

were received

from the licensee.

The licensee

did not identify as proprietary

any of

the material

provide to or reviewed

by the licensee.

Item Number

Descri tion and Reference

400/88-28-01

Violation - Failure to provide the proper monitoring

device for an individual entering

a high radiation

area,

Paragraph

2.c.

400/88-28-02

400/88-28-03

400/88-28-04

400/88-28-05

400/88-28-06

Violation - Inadequate

surveys of items released

from

a contaminated

area,

Paragraph

2.e.

IFI - Review wearing of protective clothing in

contaminatied

areas,

Paragraph

2.e.

IFI - Review documentation for pre use test of the

IPM-7

and

HECM-2A personnel

contamination

monitors,

Paragraph

2.e.

Violation - Failure to follow procedure for response

checking alarming dosimeters,

Paragraph

2.e.

Violation - Failure to review

and

approve Process

Control

Program

implementing

procedures,

Paragraph

3.d.

6.

Acronyms and Initialisms